ML20210S282

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Forwards Notice of 860806 Meeting W/Util & State of VT in Bethesda,Md to Discuss Status of Mark I Containment Safety Study.Vermont Yankee Containment Safety Study & Safety Objectives Encl
ML20210S282
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 07/15/1986
From: Houston R
Office of Nuclear Reactor Regulation
To: Stello V
NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
Shared Package
ML20209C518 List:
References
FOIA-86-586 NUDOCS 8610080070
Download: ML20210S282 (32)
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Text

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tvASHINGTON, D. C. 20555

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4 ol% . . .,. . July 15,1986 )

NOTE TO: Victor Stello, Jr.

Executive Director for Operations FROM: R. Wayne Houston, Deputy Director Division of BWR Licensing

SUBJECT:

MEETING NOTICE - VERMONT YANKEE CONTAINMENT SAFETY STUDY At Mr. Denton's request and for your information I am enclosing a copy of the Vermont Yankee letter that resulted from the meeting with Governor Kunin on June 30, 1986, and a copy of a notice of a meeting-in early August at which time we will be briefed on the status of the Vermont Yankee study.

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R. Wayne Houston, Deputy Director Division of BWR Licensing

Enclosures:

As stated cc: w/o enclosures H. Denton R. Bernero D. Muller V. Rooney hMOO O 860731 CURRAN 86-586 PDR

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k d JUL 31get Docket No. 50 271 Daniel R. Muller, Director, BWR Project Directorate #2, DBL MEMORANDUM FOR:

Vernon Rooney, Project Manager, BWR PDf2, DBL FROM:

SUBJECT:

FORTHCOMING MEETING WITH VERMONT YANKEE NUCLEAR POWER CORPORATION DATE & TIME:

Wednesday, August 6, 1986 9:00 a.m.

LOCATION: Room P-110 Phillips Building Bethesda, MD PURPOSE:

Meeting with Vermont Yankee Nuclear Power Corporation (VYNPC) representatives to review status of the Vennont Yankee Mark I containment safety study which is scheduled for completion at the end of August.

Licensee State of Vermont PARTICIPANTS *: NRC P. Paul E Muller R. Lodwick V. Rooney R. Capstick L. G. Hulman W. Hodges D. Vassallo H. Abelson Original signed by Vernan Rooney, Project Manager BWR Project Directorate #2 Division of BWR Licensing cc: See next page DISTRIBUTION:_

Docket File VRooney HI,ieelson NRC PDR OELD ACRS(10)

DMuller OPA LPDR "

._..y. LGHulman -.::. 4 PD#2 M.N. File WHodges hii ' ? "J.. : ri .

DVassallo WHouston HDenton

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HATE 27/ 1 /86 :

OFFICIAL RECORD COPY

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i, I k Mr. R. H. Capstick Vernont Yankee Nuclear Power Corporation -

Vermont Yankee Nuclear Power Station CC:

Mr. J. G. Weigand W. P. Murphy, Vice President &

President & Chief Executive Officer Manager of Operations Vermont Yankee N6 clear Power Corp. Verront Yankee Nuclear Pcwer Corp.

R. D. 5 Fox 169 R. D. 5, Box 169 Ferry Read Ferry Road Brattlebere, Vernont 05301 Brattleboro, Vernant 05301 Mr. Dcnald Hunter, Vice President Mr. Gerald Tarrant , Conrdssiener Vermont Yankee Nuclear Power Corp. Verment Department of Public , Service

- 1671 Worcester Road 120 State Street Framingham, Massachusetts 01701 Montpelier, Vermont 05602 New England Coalition en huclear Pollution Fill ar.d Dale Farm Public Service Board R. D. 2, Box 223 State of Vermont Putney, Vermont 05346 120 State Street Montpelier, Vermont 05602 Hr. Walter Zaluzny Chairman, Beard of Selectran Vermont Yankee Decommissionirg Post Of' ice Box 116 Alliance Vernon, Vermont 05345 Box 53 tsontpelier, Verrent 05602-0053 J. P. Pelletier, Plant Meneger Vermont Yankee Nuclear Power Corp.

Post Office Box 157 Resident Inscector Vernon, Vermont 05354 U. S. Nuclear Regulatory Commission Post Office Box 176 Raymond N. McCandless Vernon, Verrent 05354 Vermont Division of Occupatienti A Radiological Health Yernont Public Interest Administration Building Research Group, Inc.

10 Baldwin Street 43 State Street Montpelier, Vermont 05602 Montpelier, Vermont 05E02 Thomas A. Murley Honorable John J. Easton Regional Adninistrator Attorney General State of Verwent Region I Office

U. S. t'uclear Regulatory Corrission 109 State Street 631 Park Avenue Montpelier Vermont 05602

.' King of Prussia, Per.nsylvenia Ic406 John A. P.itscher, Esquire 2

Ropes & Gray 225 Franklin Street .

Boston, Massachusetts 02110

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/kg . 6, /ffG VERMONT YANKEE CONTAINMENT SAFETY STUDY

1. PURPOSEANDOBJECTIVEOFEACHTASKINCLUDifiGPRELIMil@RYRESULTS MARK I DESIGN REVIEW o WASH IL00 1 DESIGN COMPARIS0N o KEY DESIGN DIFFERENCES o VERMONT YANKEE DESIGN

SUMMARY

VERMONT YANKEE CONTAINMENT CAPABILITY o VERMONT YANKEE APPROACH o STUDIES AND RESOURCES USED o ACCIDENTS AND TRANSIENTS STUDIED o RESULTS UTILIZING KEY DIFFERENCES AT VERMONT YANKEE o DEFINITION OF "90%"

o VY CONTAINMENT CONDITIONAL FAILURE PP0BABILITY CURRENT MARK 1 ISSUES o DEFINITION AND TECHNICAL STATUS OF 5 ISSUES o APPLICABILITY AND NEED FOR FURTHER STUDY o ACTIVE INDUSTRY AND NRC EFFORTS

l. STATUS RESULTS 10 DATE

- SCHEDULE FOR COMPLETION f

^D

e g CONTAINMENT SAFETY STUDY

, DESIGN AND OPERATIONAL FEATURES COMPARISON Design / Operational Data Vermont Yankee (WASH-1400)

Gen ral Plant Data Plcnt Type CE BWR GE BWR Containment Type Mark I Pressure Suppression Mark I Pressure Suppression R;tcd Thermal Power, NWt 1,593 3,293 R;tcd Core Flow, Ib/hr 48.0 x 10E6 102.5 x 10E6 R tcd Steam Flow, Ib/hr 6.43 x 10E6 13.381 x 10E6 R,cetor Data Inside Height, ft-in 63 - 1.5 72 - 11 Innide Diameter, in. 205 251 C*ntainment Data Intcenal Design Pressure, psig 56 56 Drywell Data Cylinder Diameter, ft. 33 38.5 62 67 Spherical Diameter,3 ft.

Frce Air volume, ft 134,000 159,000 T*rus Data M3jor Diameter, ft. 98 111.5 Minor Diameter, ft. 27.66 31 W:ter Volume, min / max., ft3 68,000/70,000 123,000 Fr:e Air Volume, ft3 114,200/112,200 119,000 Vent Pipes Data Culber 8 8 Internal Diameter, ft. 6.75 6.75 Downcomer Pipe Data Nu;ber 96 96 Internal Diameter, f t. 2 2 Submergence, ft. (nominal) 4 4 Nutber of WW-DW Vacuum 10 12 Breakers

. Design / Operational Data Vermont Yankee (WASH-1400)

Secondary Containment Data fro 9 Air Volume ,

2,120,000 2,400,000 Mitigation Systems Design Data HPCI System Number of Trains or Subsystems 1 1 Number of Pumps / Train 1 1 Design Flow / Train 4,250 gpm at 1,120 to 150 psid 5,000 gpm at 1,100 to 150 psig Elcctrical Power DC only (turbine-driven) DC only (turbine-driven)

RCIC System tumber of Trains or Subsystems 1 1 Number of Pumps / Train 1 1 Design Flow / Train 400 600 gpm

, Elcctrical Power DC only (turbine-driven) DC only (turbine-driven)

RHR System Number of Trains or Subsystems 2 2 Number of Pumps / Train 2 2 Number of HXs/ Train 1 2 Design Flow / Train 14,400 gpm 20,000 spm Electrical Power Emergency ac and de Emergency ac and de Source of Water Torus Torus Service water (river) Service water Ultimate backup (Diesel fire pump cross-tie capability)

Emergency Diesel Generator Systems Number of Emergency Buses 2 4 per unit Number of Emergency Diesels 2* 4* - Shared between two units

  • Single-Unit Site - Requires *Two-Unit Site - Requires two one out of two emergency out of four emergency diesels diesels for safe shutdown. for safe shutdown.

'O VERMONT YANKEE - WASH 1400 KEY DESIGN DIFFERENCES RATIO DESIGN PARAMETER VERMONT YANKEE /WAS4 1400 RATED THERMAL POWER .48 DRYWELL VOLUME / POWER 1.75 l

TORUS WATER VOLUME / POWER 1.19 TORUS AIR VOLUME / POWER 1.96 HPCI PUMPING CAPACITY / POWER 1.77 RCIC PUKPING CAPACITY / POWER 1.40

VERMONT YANKEE DESIGN

SUMMARY

o SMALL REACTOR PLANT (NSSS) IN LARGE MARK I CONTAINKENT o ENGINEERED SAFETY FEATURES CAPACITY o RESIDUAL HEAT REMOVAL CAPABILITY o ELECTRIC DRIVEN MAIN FEED PUMPS o DC SYSTEM CAPACITY & DIVERSITY

- 8 HOUR BATTERY RATING

- APPENDIX R BATTERIES

- SPECIAL PURPOSE BATTERIES

- UNINTERRUPTIBLE POWER SUPPLIES o AC SYSTEM RELIABILITY & DIVERSITY

- VERNON HYDR 0 TIE LINE

- NORTHEAST GRID RELIABILITY

- SEPARATE HIGH LINE RIGHT OF WAYS o DIESEL FIRE PUMP CROSSTIE 10 RHR SYSTEM

P STUDY APPROACH

~

o IDENTIFY THE DOMINANT ACCIDENT SEQUENCES WHICH CAN LEAD 10 SEVERE ACCIDENTS AT VERMONT YANKEE.

o QUANTIFY THE DOMINANT SEQUENCES USING A REFERENCE MARK l MODIFIED FOR THE UNIQUE VERMONT YANKEE FEATURES.

o DEVELOP THE CONTAINMENT EVENT TREE TO DISPLAY THE PATHWAYS 10 SAFE MITIGATION AND POTENTIAL RADIONUCLIDE RELEASE TO THE ENVIRONMENT.

o OUANTIFY THE CONTAINMENT EVENT TREE USING AVAILABLE ESTIMATES OF MITIGATION RELIABILITY AND STANDARD MODELS.

o CALCULATE THE CONDITIONAL FAILURE PROBABILITY OF CONTAINMENT.

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STUDIES & RESOURCES

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STUDIES 4

o WASH 1L100 o IDCOR

! - IPE METHODOLOGY

- TECHNICAL

SUMMARY

REPORT

- TASK REPORT DOCUMENTS o DRAFT BWR MARK 1 PSA's o SEVERE ACCIDENT SEQUENCE ANALYSIS PROGRAM RESOURCES o YANKEE ATOMIC ELECTRIC COMPANY o DELIAN CORPORATION i

o FAUSKE AND ASSOCIATES o RISK MANAGEMENT ASSOCIATES i

o GENERAL ELECTRIC COMPANY i o PICKARD LOWE & GARRICK, INC.

5

BWR DOMINANT ACCIDENT SEQUENCES CLASS I LOSS OF COOLANT MAKEUP 1

lA HIGH PRESSURE-IB stall 0N BLACK 0UT IC ATWS I ID LOW PRESSURE lE LOSS OF DC CLASS 11 LOSS OF CONTAINMENT HEAT REMOVAL CLASS Ill LOCA CLASS IV ATWS fE Y,

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S BWR DOMINANT ACCIDENT SEQUENCES CLASS I LOSS OF COOLANT MAKEUP IA HIGH PRESSURE IB STATION BLACK 0UT IC ATWS ID LOW PRESSURE IE LOSS OF DC CLASS 11 LOSS OF CONTAINMENT HEAT REMOVAL CLASS Ill LOCA CLASS IV ATWS Y,b 56

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VERMONT YANKEE RESULTS DOMINANT SEQUENCES STAT 10r BLACK 0UT KEY DIFFERENCES: O o DIESEL FIRE PUMP CROSSTIE o VERNON HYDRO TIE LINE - M*

o DC SYSTEM CAPABILITY o DIESEL GENERATOR RELIABILITY o NEW ENGLAND GRID STABILITY AIEt KEY DIFFERENCES:

o CONTAINMENT SIZE VS. POWER o EXISTING ATWS MODIFICATIONS AND COMMITMENTS o LARGE RHR SYSTEM HEAT REMOVAL CAPACITY o MSIV REOPEN CAPABILITY k

o EMERGENCY OPERATING PROCEDURES

DEFINITION OF "90%"

WASH-1400 ASSUMPTIONS 1

l o ALL SEQUENCES (MELT /NON-MELT) ARE ASSOCIA1ED WITH CONTAINMENT FAILURE 0 CORE MELT RELEASE CATEGORIES

1. STEAM EXPLOSION
2. OVERPRESSURE FAILURE DIRECT TO ATMOSPHERE
3. OVERPRESSURE FAILURE THROUGH REACTOR BUILDING
4. ISOLATION FAILURE l

0 RELEASE CATEGORIES 2 & 3 DOMINATE RISK o SEQUENCES WHICH CONTRIBUTE COMPRISE 90%

ATWS LOSS OF RESIDUAL CORE HEAT REMOVAL

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CURREN1 STUDY o DIFFERENT DOMINANT SEQUENCES IDENTIFIED o KNOWLEDGE OF CONTAINMENT FAILURE PHENOMENA AND CRITERIA HAS EVOLVED

CONTAINMENT CONDITIONAL FAILURE PROBABILITY o CALCULATE THE CONTRIBUTION OF EACH ACCIDENI CLASS 10 101AL LIKELlH00D OF CORE MELT o DEIERMINE THE CONDITIONAL FAILURE PROBABilllY OF CONIAINMENT FOR EACH ACCIDENT CLASS o CALCULATE THE WEIGHTED AVERAGE OF PRIMARY CONTAINMENI FAILURL PROBABILITY l

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CONDITIONAL CONIAINMENT FAILURE PROBABILITY 00ANTIFICA110N DOMINANT FRACTIONAL CONTAINMENT CONTRIBUTION ACCIDENT CONTRIBUTION FAILURE TO WEIGHTED SEQUENCE (CORE MELT) PROBABILITY AVERAGE LOSS OF C0LANT MAKEUP A A AxA LOSS OF CONTAINMENT HEAT REMOVAL s B BxB LOCA c C cxC ATWS o D oxD n

1.0 {(

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) = CFFP WHERE CCFP = CONDITIONAL CONTAINMENT FAILURE PROBABILITY

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e I.O CURRENT TECHNICAL ISSUES

1. HYDROGEN CONTROL
2. DRYWELL SPRAYS
3. CONTAINMENT PRESSURE CONTROL
4. CORE DEBRIS CONTROL i
5. TRAINING AND PROCEDURES

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HYDR 0 GEN CONTROL OBJECTIVE: PREVENT HYDR 0 GEN C0fiBUST10N CAUSED FAILURE SVGGESTIONS:

OXYGEN CONTROL INERT TO START CONTROL INGRESS OF OXYGEN ,

ISSUES IDENTIFIED BY NRC:

WHEN AND HOW LONG NOT INERTED PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS

1. CONTAINMENT INERTED
2. ELECTRIC POWER NOT REQUIRED TO MAINTAIN INERT
3. TECH. SPECS. CONTROL DEINERT TIME
4. PLANT SHUTDOWN IF TECH. SPEC. CANNOT BE MET t

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DRYWELL SPRAYS OBJECTIVE: SPRAY WATER 10:

1. LOWER PRESSURE
2. COOL VULNERAPLE EQUIPMENT
3. QUENCH DEBRIS
4. SCRUB AEROSOLS SUGGESTIONS:
1. SPRAY IN DRYWELL
2. INDEPENDENT BACKUP WATER SOURCES AND PUMPS HOSE CONNECTIONS USE OF FIREMAINS ISSUES IDENTIFIED BY NRC:

4

1. RISK 0F IMPLOSION
2. RISK OF HYDROGEN COMBUSTION AFTER STEAM CONDENSATION
3. MANUAL ACTIONS AND TIMING PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:
1. EXISTING FLOW PATH FROM DIESEL FIRE PUMPS TO VESSEL OR DRYWELL SPRAYS (2500 GPM)
2. AVAILABLE FOLLOWING STATION BLACK 0UT
3. IF EMERG DIESEL AVAILABLE, SYSTEM OPERABLE FROM OUTSIDE REACTOR BUILDING
4. EXISTING OPERATING PROCEDURE DESCRIBES USE OF SYSTEM

CONTAINMENT - PRESSURE CONTROL OBJECTIVES: 1. AVERT UNCONTROLLED OVERPRESSURE FAILURE

2. CONTROL RELEASE PATH (SCRUBBING)

SUGGESTIONS:

1. SUBSTANTIAL CAPABILITY TO VENT WETWELL
2. REMOTE / RELIABLE CONTROL OF VENT VALVE
3. ABILITY TO RECLOSE VENT ISSUES IDENTIFIED BY NRC:
1. DELIBERATE RELEASE OF RADIDACTIVITY
2. WHAT CONSTITUTES REMOTE / RELIABLE CONTROL?
3. IS VENTING TO SECONDARY CONTAINMENT ACCEPTABLE?
4. WHAT IS APPROPRIATE ACTION PRESSURE?

PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:

1. EIGHTEEN INCH ATMOSPHERIC CONTROL SYSTEM VENTS
2. THREE INCH ATMOSPHERIC CONTROL SYSTEM VENT
3. TWENTY AND EIGHTEEN INCH NITROGEN PURGE LINES
4. SIX INCH NITROGEN PURGE LINES
5. ONE INCH NITROGEN CAD LINE - EXISTING PROCEDURE

CORE DEBRIS CONTROL OBJECTIVE: REDUCE LIKEllH00D OF FAILURE BY DIRECT CONTACT OF CORE DEBRIS WITH DRYWELL WALL.

SVGGESTIONS:

1. USE PRACTICAL DEBRIS RETARDING BARRIERS
2. CONSERVE SUPPRESSION P0OL WATER AS A QUENCHING P00L ISSUE IDENTIFIED BY NRC: WHAT IS PRACTICAL?

PRESENT VY CAPABILITY RELATIVE TO PROPOSED RE0VIREMENTS:

1. SMALL CORE DEBRIS VOLUME COMPARED TO PREVIOUS STUDIES
2. DRYWELL SUMPS COMBINED VOLUME APPR0XIMATELY 200 FT3
3. > 1200 FT*'DRYWELL FLOOR SURFACE AREA
4. DOWNCOMERS APPR0XIMATELY ONE FOOT AB0VE DRYWELL FLOUR

TRAINING AND PROCEDURES OBJECTIVE: ENSURE OPERATORS ARE READY TO USE PLANT FEATURES TO BEST ADVANTAGE IN SEVERE ACCIDENTS SUGGESTIONS:

1. CLEAR SYMPT 0M BASED STRATEGIES (INTEGRATED)
2. REMOVAL OF UNNECESSARY INHIBITIONS
3. TRAINING ISSUES IDENTIFIED BY NRC:
1. COMPETING SAFETY REQUIREMENTS
2. DEGREE OF TRAINING PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:
1. REV. 3 0F EMERGENCY PROCEDURE GUIDELINES (EPG's) IMPLEMENTED
2. ACTIVE PARTICIPATION IN REV. 4 DEVELOPMENT
3. VY PLANT SPECIFIC SIMULATOR COMPLETED IN 1985
4. EXTENSIVE OPERATOR TRAINING PROGRAM INCLUDES SEVERE ACCIDENT PREVENTION AND MITIGATION

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VERMONT YANKEE NUCLEAR POWER CORPORATION

. FVY86-69 RD 5 Box 169. Ferry Road Brattleboro. VT 05301 ,, o ENGINEERING OFFICE 1671 WORCESTER ROAD

. FRAMINGHAM. MASSACHUSETTS 01701

  • TELEPHOfeE 617-472-8100 August 1, 1986 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Office of Nuclear Reactor Regulation Daniel R. Muller Operating Reactors Branch No. 2 Division of Licensing

References:

a) License No. DPR-28 (Docket No. 50-271) b) Letter, J.G. Weigand to Harold Denton, dated 6/30/86

Dear Sir:

Subject:

Vermont Yankee Containment Safety Study Interim Progress Report By letter, Reference b), Vermont Yankee committed to perform a study of Vermont Yankee's Mark I containment to address enhancements to containment capa-bility which were not analyzed for the WASH-1400 reference plant, assess a best estimate Vermont Yankee specific conditional containment failure probability, and address the five elements presented by Robert Bernero at the June 16, 1986 meeting held in Bethesda.

To document our progress to date, enclosed please find a copy of our study plan, with task completion percentages noted (Attachment 1). Additionally, we have enclosed a copy of our presentation outline for the August 6, 1986 progress meeting in Bethesda (Attachment 2).

We look forward to this upcoming opportunity to meet with you and discuss progress on the Vermont Yankee Containment Safety Study.

Very truly yours, ONT YANKEE NUCLEAR POWER CORPORATION GA G Richard Lodwick Operations Support Manager cc: W. Houston V. Rooney State of Vermont

  • ATTACHMENT 1 e

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[ ass sa sue 3 YN" -i?ll".%N STATUS AS OF 08/01/86 l _ . . _

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ATTACHMENT 2 l

9 VERMONT YANKEE CONTAINMENT SAFETY STUDY

1. PURPOSE AND OBJECTIVE OF EACH TASK INCLUDING PRELIMINARY RESULTS MARK I DESIGN REVIEW o WASH 1400 DESIGN COMPARISON o VERMONT YANKEE DESIGN

SUMMARY

VERMONT YANKEE CONTAINMENT CAPABILITY o VERMONT YANKEE APPROACH o STUDIES AND RESOURCES USED o ACCIDENTS AND TRANSIENTS STUDIED o RESULTS UTILIZING KEY DIFFERENCES AT VERMONT YANKEE o DEFINITION OF "90%"

o VY CONTAINMENT CONDITIONAL FAILURE PROBABILITY CURRENT MARK I ISSUES o DEFINITION AND TECHNICAL STATUS OF 5 ISSUES o APPLICABILITY AND NEED FOR FURTHER STUDY o ACTIVE INDUSTRY AND NRC EFFORTS II. STATUS RESULTS TO DATE

- SCHEDULE FOR COMPLETION

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~US ak-SAFETY OBJECTIVES -r#<-

S . 6//o yp n/cs S/O e THE LIKELIHOOD OF SEVERE ACCIDENT (CORE DAMAGE OR CORE MELT) SHOULD BE VERY LOW AND e IF A SEVERE ACCIDENT OCCURS THERE SHOULD BE SUBSTANTIAL ASSURANCE THAT THE CONTAINMENT WILL MITIGATE ITS CONSEQUENCES O

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'* CONTAINMENT ISSUES

~

e EARLY REACTORS -

LOW POWER / BIG CONTAINMENTS COULD MEET CONTAINMENT OBJECTIVE e EVOLUTION OF DESIGN MUCH HIGHER POWER FOCUS ON PREVENTION OBJECTIVE CONTAINMENT GOOD FOR FISSION PRODUCTS BUT QUESTIONS ABOUT HEAT AND GAS e REACTOR SAFETY STUDY (1975)

BIGGER REACTORS 1 PWR (SURRY) 1 BWR (PEACH BOTTOM)

BWR RESULTS INDICATED LOWER PROBABILITY BUT P00R CONTAINMENT l

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l

'5 WHAT ARE THE PROBLE.86 AND SOLUTIONS l

l e 5 ELEMENTS TO CONSIDER

' HYDR 0 GEN SPRAYS PRESSURE CORE DEBRIS TRAINING a PROCEDURES e MANY CHANGES ARE ALREADY IN PLACE e FINAL IMPROVEMENTS ARE NOW UNDER HIGH PRIORITY STUDY GENERIC ACTION WITH IDCOR AND BWROG GENERIC WORK BY NRC VERMONT YANKEE STUDY PILGRIM PROGRAM 9

, SEVERE ACCIDENT SAFETY IN BOILING WATER REACTORS '

WITH MARK I CONTAllMENT As the name indicates, a boiling wated reactor (8WR) is a reactor in which the water fed to the reactor core boils right there in the reactor vessel and then passes as steam directly out to the turbine generator where its energy is converted to electricity. The exhausted steam, after condensation, is returned to the reactor as feedwater. Figure I shows a simple schematic of a i BWR plant. The reactor is enclosed in a special containment structure. The feedwater enters and the steam leaves this containment structure through multiple, large diameter pipes equipped with redundant valves which can be closed in an emergency. In the pressure suppression containment which is used in all large U.S. BWRs, a very large quantity of water, up to one million gallons, is stored in a special compartment of the containment called the suppression pool. Many auxiliary and emergency cooling systems are provided to pump cooling water into the reactor and to cool the containment atmosphere and its suppression pool. If a pipe breaks by accident, the containment closes to isolate the reactor in the containment and many cooling systems are called into play to cool the reactor and the suppression pool, removing the stored energy and heat generated by radioactive decay.

Thus, the BWR is an open system removing large quantities of energy to nearby equipment which, in emergencies, converts to a closed system, basically relying on external cooling of the containment to remove the bottled-up energy. The most common type of pressure suppression containment in the U.S.

is the Mark I type shown in Figure 2, which is used in the 24 U.S. BWRs listed in Table 1. The reactor is contained in the drywell portion of the containment, shaped like an electric light bulb standing upside down. The suppression pool partially fills a toroidal shell around the base of the

" bulb" and a series of ducts is installed to guide steam and other releases into the suppression pool which quenches the steam and also absorbs much of the radioactive material (except gases).

l-

" Severe accidents" is the term most commonly used to describe accidents in ,

which the reactor core is severely damaged. As happened at Three Mile Island, prolonged loss of core cooling can allow the heat of radioactive decay in the

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core to build up to the point that the fuel begins to disintegrate, the zirconium metal cladding melts or reac'ts with residual steam to form combustible hydrogen, and even the ceramic uranium oxide fuel pellets can 4

melt. A great deal of attention is being given to understanding the behavior of reactors and their containments in severe accidents, especially since the Three Mile Island accident. The objectives are to ensure that the likelihood of core melt accidents is very low and that, should one occur, there is substantial assurance that the containment will mitigate its consequences.

The severe accident behavior of a BWR with a Mark I containment, the Peach Bottom plant, was assessed in the Reactor Safety Study (WASH-1400 or NUREG-75/014) which was published in 1975. That study indicated a relatively low overall risk for the BWR, principally due to its ability to prevent core melt. The containment was estimated to provide very little mitigation of core melt consequences because the buildup of pressure under accident conditions would be a direct cause of containment failure unless adequate cooling was preserved. Consistent with operating procedures in place in 1975, the Study assumed little effort by the reactor operators which might effectively preserve the containment's integrity.

The situation, more than ten years later, is different and still changing for the better. It is recognized today that molten core material melting into the

. ground through the thick containment base is not the principal threat; rather, it is an atmospheric release of radioactive material which is the principal threat. The principal factors which can cause containment failure with '

atmospheric release are hydrogen ignition, gas overpressure buildup to rupture, and direct attack of the drywell by core melt debris. The general situation for each of these is summarized as follows:

Hydrogen Ignition Recognizing Shat combustible hydrogen can be generated and released in severe accidents, all Mark I containments now are purged and filled with inert nitrogen .

gas during operation so that even if hydrogen gas is formed it has insufficient oxygen available to support combustion. Remaining questions in this area relate to how long the containment may be without this inert atmosphere in order to

'i * .. . _ . . . .

permit inspections, and how air might leak in or hydrogen leak out to nearby rooms under accident conditions.

Overpressure Failure Careful analysis indicates that a typical Mark I containment can withstand pressures of more than twice the design pressure without rupture.

Nevertheless, severe accidents in the extreme can generate such pressures and cause containment rupture. Overpressure damage control procedures have been developed for pressure suppression containments and are already in place for operator use. With these procedures the containment remains closed for most accident conditions; but, if overpressure failure threater.s, large vent valves above the suppression pool chamber are opened so that the excess pressure is released gradually by bubbling the releases through the pool, forming a filtered vent containment system. With this path assured, virtually nothing but the noble gases are released. The radioactive noble gases pose a modest exposure threat offsite only in the area very close to the plant. A number of questions are being pursued in this area. All the plants have suitably large vent valves and ducts but they vary one to another in the ability to open these valves under accident conditions. The valves are designed for highly reliable closure, not opening. Consideration is being given to modifying valve controls. In addition, the vent ductwork downstream of the valves may warrant modification. In most plants it is fairly light gauge ductwork and might be breached in accident venting. If so, consideration is being given to the effects of secondary release of radioactive gas, hydrogen, and perhaps steam into the reactor building.

Direct Attack The core melt debris, since it has melted through the' reactor vessel into the drywell say, by direct radiation of heat, cause failure of connections in the drywell shell; or the debris, if sufficiently fluid, may flow out to the wall andmeltthrNughthesteel. The Mark I containments have one or more spray systems in the drywell which are able to spray water along the walls and onto ,

the floor of the drywell inhibiting direct attack. Concerns in this area are in three general areas: core debris modeling, shell and concrete attack modeling, and spray reliability. In the first area, it is recognized that a molten reactor core, to melt through the bottom of a BWR, must dissolve a very

S 4-large amount of inert metal in the low're reactor vessel, probably diluting the core melt. The key question is whether the melt would come out moving sluggishly like Hawaiian volcano lava or as a hot free flowing liquid. The latter is the more threatening condition.

If core melt debris reaches the concrete floor and steel shell of the wall, it is important to understand that the path to the outside that might be opened bypasses the beneficial scrubbing of radioactive material passing through the pool.

As noted earlier all these plants have drywell spray systems, but they are designed as a secondary mode of operation for a reactor safety system. Strong consideration is being given to enabling hookup of these systems to fire protection systems so that spray capability is almost always available.

Substantially different emergency operating procedures and training were put in place at all reactors after the Three Mile Island accident; further improvements in these procedures are still being made. For the Mark I containments both industry and NRC studies are being used to identify the best combined strategy for procedures and perhaps some changes in equipment such as alternate vent paths, cr improved valve operability. The Mark I studie: are being given highest priority by the NRC staff and the industry. The expectation is that, with modest improvements of this type, one can achieve substantial assurance of core melt consequences mitigation by a Mark I containment.

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TABLE 1 BOILING WATER REACTORS WITH MARK I CONTAINMENTS .-

LICENSED OPERATING PLAN 4, POWER LICENSE NAME LEVEL DATE COUNTY STATE UTILITY '

BROWNS FERRY 1 3293 12/20/73 LIMESTONE COUNTY Al TVA BROWNS FERRY 2 ~ 3293 08/02/74 LIMESTONE COUNTY AL TVA BROWNS FERRY 3 3293 08/18/76 LIMESTONE COUNTY AL TVA BRUNSWICK 1 2436 11/12/76 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT 8RUNSWICK 2 2436 12/27/74 BRUNSWICK COUNTY NC CAROLINA POWER & LIGHT COOPER 2381 01/18/74 NENEHA COUNTY NE NEBRASKA PUBLIC POWER OISTRICT-ORESDEN 2 2527 12/22/69 GRUNDY COUNTY IL COMMONWEALTN EOISON ORESDEN 3 2527 03/02/71 GRUNDY COUNTY IL COMMONWEALTN EOISON DUANE ARNOLO 1658 02/22/74 LINN COUNTY IA IDWA ELECTRIC POWER & LIGHT l FERMI 2 3292 07/15/85 MONROE COUNTY MI DETROIT EDISON j FITZPATRICK 2436 10/17/74 OSWEGO COUNTY NY POWER AUTHORITY OF STATE OF NY HATCH 1 2436 10/13/74 APPLING COUNTY GA GEORGIA POWER l HATCH 2 2436 06/13/78 APPLING COUNTY GA GEORGIA POWEK -

! HOPE CREEK 1 3293 04/11/86 SALEM COUNTY NJ PUBLIC SERVICE ELECTRIC & GAS I MILLSTONE 1 2011 10/16/70 NEW LONDON CT NORTNEAST NUCLEAR ENERGY MONTICELLO 1670 01/19/71 WRIGHT COUNTY MN NORTHERN STATES POWER:

I NINE MILE POINT 1 1850 08/22/69 OSWEGO COUNTY NY NIAGARA MOHAWK POWER i

OYSTER CREEK 1 1930 08/01/69 OCEAN COUNTY NJ GPU NUCLEAR CORP PEACH BOTTOM 2 3293 12/14/73 YORK COUNTY PA PHILADELPHIA ELECTRIC i

PEACH BOTTOM 3 .3293 07/02/74 YORK COUNTY PA PHILADELPHIA ELECTRIC PILGRIM 1998 06/08/72 PLYMOUTH COUNTY MA BOSTON E0lSON QUA0 CITIES 1 2511 12/14/72 ROCK ISLAND COUNTY IL COMIONWEALTH EDIS0N 2511 ~12/14/72 ROCK ISLAND COUNTY IL COMMONWEALTH EDIS0N QUAO CITIES 2 VERMONT YANKEE 1593 02/02/73 WINDHAM COUNTY VT VERMONT YANKEE NUCLEAR POWER I

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, VERMONT YANKEE NUCLEAR POWER CORPORATION

. FVY86-69 RD 5. Box 169, Ferry Road, Brattleboro, VT 05301 ,,,,,,o y ENGINEERING OFFICE 1671 WORCESTER ROAD

. FR AMINGHAM, M ASS ACHUSETTS 01701

  • TELEPHONE 617-872-4100 August 1, 1986 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Office of Nuclear Reactor Regulation Daniel R. Muller Operating Reactors Branch No. 2 Division of Licensing

References:

a) License No. OPR-28 (Docket No. 50-271) b) Letter, J.G. Weigand to Harold Denton, dated 6/30/86

Dear Sir:

Subject:

Vermont Yankee Containment Safety Study Interim Progress Report By letter, Reference b), Vermont Yankee committed to perform a study of Vermont Yankee's Mark I containment to address enhancements to containment capa-bility which were not analyzed for tFe WASH-1400 reference plant, assess a best estimate Vermont Yankee specific conditional containment failure probability, and address the five elements presented by Robert Bernero at the June 16, 1986 meeting held in Bethesda.

To document our progress to date, enclosed please find a copy of our study plan, with task completion percentages noted (Attachment 1). Additionally, we have enclosed a copy of car presentation outline for the August 6, 1986 progress meeting in Bethesda (Attachment 2).

We look forward to this upcoming opportunity to meet with you and discuss progress on the Vermont Yankee Containment Safety Study.

Very truly yours, ONT YANKEE NUCLEAR POWER CORPORATION l 01 C Richard J Lodwick Operations Support Manager cc: W. Houston V. Rooney State of Vermont

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VERMONT YANKEE CONTAINMENT SAFETY STUDY

1. PURPOSE AND OBJECTIVE OF EACH TASK INCLUDING PRELIMINARY RESULTS MARK I DESIGN REVIEW o WASH 1400 DESIGN COMPARIS0N o VERMONT YANKEE DESIGN

SUMMARY

VERMONT YANKEE CONTAINMENT CAPABILITY o VERMONT YANKEE APPROACH o STUDIES AND RESOURCES USED o ACCIDENTS AND TRANSIENTS STUDIED o RESULTS UTILIZING KEY DIFFERENCES AT VERMONT YANKEE o DEFINITION OF "90%"

o VY CONTAINMENT CONDITIONAL FAILURE PROBASILITY CURRENT MARK I ISSUES o DEFINITION AND TECHNICAL STATUS OF 5 ISSUES o APPLICABILITY AND NEED FOR FURTHER STUDY o ACTIVE INDUSTRY AND NRC EFFORTS II. STATUS

- RESULTS TO DATE

- SCHEDULE FOR COMPLETION

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